Growing interest has recently been taken in the development of fourth generation nuclear reactors, which alleviate international resistance to nuclear proliferation (i.e. the spread of nuclear weapons). Among them, it has been reported that a sodium cooled fast reactor has entered the step just before commercialization, because the technology thereof has been considerably developed.
Thus, in several countries, including the USA, Japan and Russia, nuclear reactor design concepts that obviate a blanket and use a single-enrichment nuclear fuel in order to alleviate resistance to the spread of liquid-metal nuclear reactors, have been suggested.
Each blanket contains depleted uranium or natural uranium, and is characterized by the production of nuclear-grade plutonium when loaded into a reactor core.
However, since the blanket increases the breeding ratio of nuclear fuel, the obviation of the blanket has an advantage in that it basically blocks the production of nuclear-grade plutonium, thus alleviating resistance based on fears of nuclear proliferation, but has a disadvantage in that it has a lower breeding ratio.
The use of the single-enrichment nuclear fuel maintains a constant breeding ratio in each reactor core region, and thus limits the frequency of changes in power distribution over time.
Accordingly, the single-enrichment nuclear fuel facilitates the design of an orifice required for flow distribution.
In Korea, a sodium-cooled reactor of 600 MWe has already been designed, thus both obviating the blankets and using the single-enrichment nuclear fuel. In a core of the sodium-cooled reactor, in order to flatten the power distribution using the single-enrichment nuclear fuel, a non-nuclear fuel rod, such as a B4C rod, a ZrH2 rod, or a vacancy rod, has been used for a nuclear fuel assembly.
However, the design of such a nuclear fuel assembly is complicated compared to that of a known nuclear fuel assembly. When the B4C or ZrH2 rod is irradiated at high temperature, the soundness of this non-nuclear fuel rod containing this compound becomes a problem.
Particularly, in the case of the ZrH2 rod, because hydrogen is emitted at a high temperature of 550° C., the soundness of cladding has become a known problem.
The general concept of a high-capacity sodium-cooled fast reactor is based on the use of the blankets and variation of fuel enrichment in each reactor core region in order to flatten power distribution. In the case where the single-enrichment nuclear fuel is used, a method of properly disposing the blankets in the middle of the reactor core as well as in the reactor core has been used.
In a lead-cooled fast reactor, BREST, which has recently been proposed in Russia, a method of varying the outer diameter of the nuclear fuel rod in each reactor core region instead of avoiding variation in fuel enrichment has been adopted. Further, an attempt has been made to alleviate resistance to nuclear proliferation by obviating the blanket.
The concept of the BREST is based on the use of lead as a coolant and nitride as a nuclear fuel.
However, with this method, it is difficult to maintain the size of the nuclear fuel assembly consistent due to the difference between the outer diameters of the nuclear fuel rods, when applied to the case of using a sodium-cooled reactor, particularly, a duct and a wire wrap.
The Japan Nuclear Cycle Development Institute (JNC) has proposed a method of maintaining both the outer diameter of the nuclear fuel rod and the thickness of the cladding uniform and of varying the content of zirconium (Zr) in the nuclear fuel surrounded by the cladding in each reactor core region.
Regarding this method, however, the predominant opinion is that the performance and production of the nuclear fuel become a problem.
The reason is as follows. According to results released from the US Argonne National Laboratory, the optimal content of Zr is generally 10 wt %, and when the content of Zr is less than 10 wt %, eutectic and melting temperatures of a metal fuel do not become sufficiently high, and re-distribution of elements may create a region where the content of Zr is abruptly reduced. In contrast, when the content of Zr is more than 10 wt %, the melting temperature of a fuel core becomes higher than that of a quartz tube when the nuclear fuel is produced.
JNC has also proposed a method of simultaneously varying the content of Zr in each reactor core region and adjusting the smear density of the metal fuel.
However, according to a study on the smear density of the metal fuel, reported by US Argonne National Laboratory, the optimal smear density of the metal fuel is 75% TD, and when it is higher than this value, the soundness of the nuclear fuel become a problem.
According to the design concepts of an encapsulated nuclear heat source (ENHS), a small-size ultra-long cycle lead-cooled fast reactor, which is being studied under the control of California State University at Berkeley in the USA, an attempt has been made to maintain the size and components of the nuclear fuel rod uniform in all of the reactor core regions, and to adjust the power distribution by increasing the area of a non-nuclear fuel region in the middle of the fuel core.
In this case, however, it is found that the area of the non-nuclear fuel region in the middle of the fuel core is increased such that the peak power factor is maintained less than 1.5, and thus the breeding ratio is considerably reduced.
According to the ENHS design concepts, another method of maintaining the design specification and components of each nuclear fuel rod in each reactor core region uniform and of varying only the distance between the nuclear fuel rods has been released.
This method, however, is suitable for a small-size reactor core that makes no use of a duct and a wire wrap, but not for a large-size reactor core, which uses the duct and the wire wrap.